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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Estimates of Dose Rates from the 

Plutonium-bearing Fuel Pellets Fabricated

at Carborundum

White Paper

National Institute for Occupational 

Safety and Health

August 3, 2018

Joseph Guido

Oak Ridge Associated Universities Team

Reviewed by Thomas P. Tomes

Division of Compensation Analysis and Support
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

Page 2 of 9

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution.

1.0 INTRODUCTION

The Carborundum Company in Niagara Falls, NY, performed reactor fuel development work for 

the Atomic Energy Commission (AEC). From 1961 through 1967, Carborundum operated a 

facility for the study and fabrication of plutonium refractory materials for an AEC fuel 

development program (NIOSH 2015).

This report provides estimates of dose rates from the plutonium-bearing fuel pellets fabricated at 

the facility. The support files used in this analysis are available in ORAUT (2018).

2.0 GEOMETRY AND MATERIALS

2.1 Pellet Geometry

The radiation source for these dose calculations is a small pellet of Pu-U carbide, having a 

density of 12.8 g/cm3 (Strasser and Stahl 1965). The carbide fuel pellets produced by 

Carborundum had a range of densities. This density (12.8 g/cm3) was mentioned in the text of 

the final report as starting material for experiments related to out-of-pile properties of mixed 

uranium-plutonium carbides (Stahl and Strasser 1963, PDF p. 14). The density chosen was 

identified as the density of fuel that had been sintered without sintering aids. This density was 

chosen to maximize the amount of fuel in a single fuel pellet. Fuel that had been sintered with a 

sintering aid (0.1 w/o Ni) had a higher density of 13.1 g/cm3 (Stahl and Strasser 1963, PDF p. 

14). Doses estimates were made for a single pellet that is a cylinder 0.2 inches in diameter and 

0.2 inches tall (Strasser and Stahl 1965). The pellets were assumed to be unclad.

2.2 Glovebox Geometry

The pellets were assumed to be housed in a glovebox three ft high, five ft long, and 3.5 ft deep 

(Saulino et al. 1962). The side walls, roof, and floor of the glovebox were 1⁄4-in. welded 

aluminum plate. The front and back walls were 1⁄4-in. safety glass. The safety glass consisted of 

two 1⁄8-in. layers of plate glass and a 15-mil inner layer of polyvinyl butyral that inhibits 

shattering of the glass.

In addition to the glovebox, the model also contained a concrete back wall, a ceiling, and a floor 

to generate albedo. A 1/16-in. thick steel panel was implemented on the ceiling and walls of the 

room.

3.0 RADIATION SOURCE TERMS

3.1 Isotopic Compositions of U and Pu

The plutonium was assumed to be weapons grade plutonium having the weight fractions shown 

in Table 1 (Strasser and Taylor 1962). The uranium was enriched uranium as shown in Table 2. 

Three isotopic compositions for plutonium were used during the course of the experiments at 

Carborundum. The isotopic composition shown in Table 1 is that of plutonium that was 
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

provided by Hanford (Strasser and Taylor 1962, PDF p. 15). The isotopic composition was 

chosen because this material had the highest weight fraction of 241Pu. Plutonium-241 decays to 

241Am, which drives the photon dose because it emits a 60 keV photon that penetrates shielding 

more readily than the lower energy photons emitted by plutonium isotopes. 

The isotopic composition of the uranium used for the carbide fuel pellets was not explicitly 

stated in the reports. An enrichment of 24% was chosen to be consistent with peer review 

calculations performed by SC&A (2016) and consistent with material known to have been used 

at Carborundum (Strasser and Taylor 1962). The calculated dose rate from the fuel pellets is not 

strongly influenced by the isotopic composition of the uranium used for the fuel pellets. 

Table 1. Isotopic Analysis of Pu. 

Nuclide Mass Fraction

238Pu 0.00

239Pu 0.907

240Pu 0.079

241Pu 0.012

242Pu 0.001

Table 2. Isotopic Analysis of U. 

Nuclide Mass Fraction

234U 0.00132

235U 0.24

238U 0.75868

3.2 Photon Source Calculations 

The photon component of the dose rates from (U0.8 Pu0.2)C is due mainly to the radioactive 

progeny of the nuclides listed in Table 1 and Table 2. The activities of the progeny radionuclides 

at five years after separation were calculated using the Radiological Toolbox 3.0.0 (Eckerman 

and Sjoreen 2013). The photon intensities and energies emitted by the radionuclides were 

obtained from Be et al. (2004) and the International Commission on Radiological Protection 

(ICRP 2008). 

3.3 Neutron Source Calculations 

The neutron source term was calculated using the SOURCES 4C computer code (Wilson et al. 

2002). The volumetric source considered for these calculations considered included neutrons 

emitted from within the volume of the source region by spontaneous fission and by alphaneutron (alpha,n) interactions.
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Input to SOURCES 4C included the desired energy structure of the neutron spectrum, the 

number density of the actinide isotopes, and the fraction of the (alpha,n) target nuclides to the 

total number density. Neutron spectra were calculated in 760 groups with a minimum energy of 

0 MeV and a maximum energy of 4.49 MeV. SOURCES 4C reports that 240Pu produces about 

73% of all spontaneous fission and (alpha,n) neutrons that form the source term.

Carbon-13 was the only target isotope considered for calculations of (alpha,n) neutrons because 

the atom fraction of 13C exceeded the atom fractions of other impurity isotopes by about a factor 

of 100. 

4.0 DOSE EQUIVALENT CALCULATIONS

4.1 Fluence to Dose Conversion Coefficients 

MCNP6TM (Pelowitz 2013) was used to calculate the fluence rates of neutrons and photons that 

entered volume elements that represented dosimeters. Two dosimeters were implemented in the 

MCNP6 input file; one at 30.48 cm from the center of the fuel pellet and one at 100 cm from the 

center of the fuel pellet. Conversion coefficients were used to translate the fluence rates to dose 

equivalent rates. The conversion coefficients were included in the MCNP6 input file and the 

conversion from fluence to dose equivalent was performed by MCNP6. 

The neutron conversion coefficients used to generate the data in Table 3 can be found in Table 

A.42 of ICRP Publication 74 (ICRP 1996). The photon data in Tables 4 and 5 were derived 

using conversion coefficients found in Tables A.1, A.21, and A.24 of ICRP (1996). 

4.2 Tallies

The neutron and photon doses were accumulated into energy bins as described by the Office of 

Compensation and Analysis (NIOSH 2007). Photon exposures were divided into three energy 

bins (< 30 keV, 30-250 keV, and >250 keV). Neutrons were divided into five energy bins (< 10 

keV, 10-100 keV, 100-2000 keV, 2-20 MeV, and >20 MeV). For both neutrons and photons, the 

upper energy bins were divided into two bins to help elucidate the energy spectra of the 

radiations. In particular, it was desired to determine if the tally energies exceeded the energy of 

the dose conversion factors. 

The tallies were accumulated into two air volumes that represented dosimeters. The air volumes 

were 5.08 cm × 5.08 cm × 0.2 cm, and the centers were located 30.48 cm (1 ft) and 100 cm from 

the center of the fuel pellet. The 30.48 cm location approximates the abdomen of the Nuclear 

Technology Services, Inc. (NTS) Bottle Mannequin Absorber (BOMAB) phantom described in 

the glovebox Technical Information Bulletin (NIOSH, 2011a). The dosimeter air volumes are 

approximately the size of a dosimeter.
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

5.0 MCNP6 RESULTS

Dose equivalent rate information for a single fuel pellet are provided in Tables 3–5. The photon 

dose rates exceed the neutron dose rates. A plot of the photon spectrum at the 30.48 cm 

dosimeter location, not shown here, indicates that the major contributor to the photon dose rate is 

the 59.5 keV photon from 241Am. The values in Tables 3–5 below are based on a single pellet. 

Section 6.0 provides a dose model based on the amount of material in process and typical 

working conditions. 

Table 3. Neutron Dose Component of Neutron Source Term Calculation.

Neutron energy 

(MeV) Hp(10) (mRem.h-1) Error H*(10) (mRem.h-1) Error

0–0.01a 8.7147E-08 0.0099 7.8582E-08 0.0099

0.01–0.1a 1.3035E-06 0.0042 1.3058E-06 0.0042

0.1–2.0a 3.3814E-04 0.0006 3.2707E-04 0.0006

2.0–20.0a 4.8368E-04 0.0005 4.6333E-04 0.0005

Totala 8.2322E-04 0.0004 7.9179E-04 0.0004

0–0.01b 6.8832E-08 0.0116 6.2225E-08 0.0116

0.01–0.1b 2.7467E-07 0.0107 2.7460E-07 0.0107

0.1–2.0b 4.3184E-05 0.0017 4.1759E-05 0.0017

2.0–20.0b 5.8541E-05 0.0014 5.6074E-05 0.0013

Totalb 1.0207E-04 0.0011 9.8169E-05 0.0011

a. Dosimeter 30.48 cm (1 ft) from center of fuel pellet.

b. Dosimeter 100 cm from center of fuel pellet.
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

Page 6 of 9 

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Table 4. Photon Dose Component of Neutron Source Term Calculation.

Photon 

Energy

(MeV)

Hp(10) 

(mRem.h-1) Error H*(10) (mRem.h-1) Error Air Kerma (mRad.h-1) Error

0.0–0.03a 1.7751E-10 0.0469 1.7435E-10 0.1456 2.2925E-10 0.0577

0.03–0.25a 1.2270E-08 0.0088 1.1433E-08 0.0088 7.8356E-09 0.009

0.25–10.0a 5.9759E-07 0.0024 5.9558E-07 0.0095 5.1669E-07 0.0024

Totala 6.1004E-07 0.0023 6.0718E-07 0.0091 5.2476E-07 0.0024

0.0–0.03b 1.3352E-11 0.1445 1.3033E-11 0.1456 1.7789E-11 0.1917

0.03–0.25b 2.9254E-09 0.0180 2.7187E-09 0.0180 1.8412E-09 0.0183

0.25–10.0b 4.7926E-08 0.0095 4.7601E-08 0.0095 4.1110E-08 0.0097

Totalb 5.0865E-08 0.0023 5.0333E-08 0.0091 4.2969E-08 0.0093

a. Dosimeter 30.48 cm (1 ft) from center of fuel pellet.

b. Dosimeter 100 cm from center of fuel pellet.

Table 5. Photon Source Term Dose Calculation.

Photon Energy (MeV) Hp(10) (mRem.h-1) H*(10) (mRem.h-1) Air Kerma (mRad.h-1) 

0.0–0.03a 5.0170E-05 4.9554E-05 6.9764E-05

0.03–0.25a 2.3594E-02 2.1704E-02 1.2798E-02

0.25–10.0a 2.4844E-03 2.3750E-03 1.8683E-03

10.0–100a 0.0000E+00 0.0000E+00 0.0000E+00

Totala 2.6129E-02 2.4129E-02 1.4736E-02

0.0–0.03b 5.7146E-06 5.6488E-06 6.5398E-06

0.03–0.25b 3.0126E-03 2.7686E-03 1.6304E-03

0.25–10.0b 3.0763E-04 2.9422E-04 2.3178E-04

10.0–100b 0.0000E+00 0.0000E+00 0.0000E+00

Totalb 3.3260E-03 3.0685E-03 1.8687E-03

a. Dosimeter 30.48 cm (1 ft) from center of fuel pellet.

b. Dosimeter 100 cm from center of fuel pellet.

Monte Carlo calculations were performed that compared the transmission of 15 keV photons 

through 1⁄4-in. plate glass and 1⁄4-in. Lucite (an alternative window material). When the dose 

location is one foot from the source, the fluence through the 1⁄4-in. equivalent plate glass is 

reduced by a factor of 3.8E-05. When the source-receptor distance was one meter, the 

transmission of 15 keV (and lower) photons through the 1⁄4-in. plate-glass-equivalent window 

was zero.
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

6.0 DOSE SCENARIOS AND ANNUAL DOSE ASSIGNMENT

Dose rates were calculated at one foot (30.48 cm) and one meter from the center of a single 

pellet (Table 6). The dose rates from a single pellet were scaled upward to account for the 

maximum amount of plutonium allowed to be processed in any batch. Consistent with the 

Technical Basis Document: Site Profiles for Atomic Weapons Employers That Worked Uranium 

Metals (NIOSH 2011b) (and following the approach used to calculate uranium exposures above), 

the following assumptions are made about a worker’s exposure conditions: 

• Operator: 50% of the workday was spent at one foot (30.48 cm) from the fuel 

object containing 525 grams of pellets (i.e., 100 grams of Pu). 

• Radiation (Rad) Production Support: 50% of the workday was spent at one meter

from the fuel object containing 525 grams of pellets (i.e., 100 grams of Pu). 

• Non-Rad Production Area: Exposure was equal to 50% of the Rad Production 

Support exposure. 

• Administrative: Exposure was equal to 10% of the Non-Rad Production Area 

exposure. 

Table 6. External Dose from 2nd Operational Period for Plutonium (1959 to 1967). 

Job Categorya

Photons 

(keV)b

<30

(rem/y)

Photons 

(keV)b 

30–250

(rem/y)

Photons 

(keV)b

>250

(rem/y)

Neutrons 

(MeV)b

<0.01 

(rem/y)

Neutrons 

(MeV)b 

0.01–0.1

(rem/y)

Neutrons 

(MeV)b 

0.1–2.0

(rem/y)

Neutrons 

(MeV)b

2.0–20

(rem/y)

Operator 0.020 8.551 0.936 N/Ac 0.001 0.129 0.183

Rad Production 

Support 0.002 1.091 0.116 N/Ac N/Ac 0.016 0.022 

Non-Rad 

Production Area 0.001 0.545 0.058 N/Ac N/Ac 0.008 0.011 

Administrative <0.001 0.055 0.006 N/Ac N/Ac N/Ac 0.001

a. The following assumptions are made about a worker's exposure conditions: Operator: 50% of the workday was 

spent at one foot (30.48 cm) from a fuel object containing 525 grams of pellets (i.e., 100 grams of Pu), Rad 

Production Support: 50% of the workday was spent at one meter from a fuel object containing 525 grams of 

pellets (i.e., 100 grams of Pu), Non-Rad Production Area: Exposure was equal to 50% of the Rad Production 

Support exposure, Administrative: Exposure was equal to 10% of the Non-Rad Production Area exposure.

b. Assigned as a constant distribution. Specific organ dose values calculated using the applicable H*10 dose 

conversion factors and ICRP 60 (1991) neutron weighting factors.

c. All dose rates were less than 0.001 rem/year. Therefore, no dose is assigned.

Photon and neutron doses are treated as a constant distribution and then multiplied by the 

applicable H*10 dose conversion factors. All photon doses are assigned as acute exposures. All 

neutron doses are assigned as chronic exposures.
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution.

7.0 REFERENCES

Be, M. M., V. Chiste, C. Dulieu, E. Browne, V. Chechev, N. Kuzmenko, R. Helmer, A. Nichols, 

E. Schonfeld, and R. Dersch, 2004, Table of Radionuclides (Volume 2 - A = 151 to 242), 

BIPM-5, Bureau International des Poids et Mesures, Paris, France. [SRDB Ref ID: 

168372]

Eckerman, K. F. and A. L. Sjoreen, 2013, Radiological Toolbox User’s Guide, NUREG/CR7166, ORNL/TM-2013/16, Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S. 

Nuclear Regulatory Commission, Washington, D.C., May. [SRDB Ref ID: 156869].

Executable copies of the Radiological Toolbox are available at: 

https://www.ornl.gov/crpk/software.

ICRP (International Commission on Radiological Protection), 1991, “1990 Recommendations of 

the International Commission on Radiological Protection, ICRP Publication 60,” Annals of 

the ICRP, volume 21, number 1–3. [SRDB Ref ID: 11517]

ICRP (International Commission on Radiological Protection), 1996, “Conversion Coefficients 

for use in Radiological Protection Against External Radiation, ICRP Publication 74,” 

Annals of the ICRP, volume 26, number 3/4. [SRDB Ref ID: 7979]

ICRP (International Commission on Radiological Protection), 2008, “Nuclear Decay Data for 

Dosimetric Calculations, ICRP Publication 107,” Annals of the ICRP, volume 38, number 

3. [SRDB Ref ID: 136400]

NIOSH (National Institute for Occupational Safety and Health), 2007, External Dose 

Reconstruction Implementation Guideline, Rev. 3, ICAS-IG-001, Office of Compensation 

and Analysis, Cincinnati, Ohio, November 21. [SRDB Ref ID: 38864]

NIOSH (National Institute for Occupational Safety and Health), 2011a, Best Estimate External 

Dose Reconstruction for Glovebox Workers, Rev. 03, Technical Information Bulletin, 

DCAS-TIB-0010, Division of Compensation Analysis and Support, Cincinnati, Ohio, 

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NIOSH (National Institute for Occupational Safety and Health), 2011b, Technical Basis 

Document: Site Profiles for Atomic Weapons Employers That Worked Uranium Metals, 

Rev. 0, Batelle-TBD-6000, Division of Compensation Analysis and Support, Cincinnati, 

Ohio, November 28. [SRDB Ref ID: 101251]

NIOSH (National Institute for Occupational Safety and Health), 2015, SEC Petition Evaluation 

Report: Petition SEC-00223, Rev. 1, Cincinnati, Ohio, June 3. [SRDB Ref ID: 172457]

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172546]
White Paper Estimates of Dose Rates from the Plutonium-bearing Fuel Pellets 

Fabricated at Carborundum

August 3, 2018

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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Pelowitz, D. B. (ed.), 2013, MCNP6TM User’s Manual, Version 1.0, Rev. 0, LA-CP-13-00634, 

Los Alamos National Laboratory, Los Alamos, New Mexico, May. The manual is 

provided in the MCNP6 release package available from the Radiation Shielding 

Information Computational Center (RSICC). MCNP6 is distributed by RSICC at: 

https://rsicc.ornl.gov/Default.aspx.

Saulino, F. A., J. C. Andersen, and K. M. Taylor, 1962, “Research Facility for the Synthesis and 

Fabrication of Refractory Plutonium Materials,” 10th Hot Lab Proceedings, ANS, pp. 277–

286, November. [SRDB Ref ID: 56948] 

SC&A, Inc., 2016, Review of the Carborundum Special Exposure Cohort (SEC) Petition-00223 

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Anigstein, Vienna, Virginia, January 27. [SRDB Ref ID: 169330] 

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Company, Niagara Falls, New York, December 6. [SRDB Ref ID: 65005] 

Strasser, A. and D. Stahl, 1965, Carbide Fuel Development Final Report, Period of May 15, 

1959 to October 15, 1965, Vol. 1, UNC-5134, United Nuclear Corporation, Elmsford, New 

York, The Carborundum Company, Niagara Falls, New York, October 15. [SRDB Ref ID: 

53617] 

Strasser, A. and K. Taylor, 1962, Carbide Fuel Development Progress Report, Period of 

September 1, 1961 to December 31, 1961, Vol. 1, UNC-5003, United Nuclear Corporation, 

Elmsford, New York, The Carborundum Company, Niagara Falls, New York, February 9. 

[SRDB Ref ID: 61996] 

Wilson W. B., R. T. Perry, E. F. Shores, W. S. Charlton, T. A. Parish, G. P. Estes, T. H. Brown, 

E. D. Authur, M. Bozoian, T. R. England, D. G. Madland, and J. E. Stewart, 2002,

SOURCES 4C: A Code for Calculating ([alpha], n), Spontaneous Fission, and Delayed

Neutron Sources and Spectra, LA-UR-02-1839, Los Alamos National Laboratory, Los

Alamos, NM. [SRDB Ref ID: 172456]. Sources 4C is distributed by the Radiation

Shielding Information Computational Center at: https://rsicc.ornl.gov/Default.aspx.